This invention relates to a method for flattening the curve which is representative of the evolution of heat within a nuclear reactor core in the radial direction and in the axial direction of said core. The invention also relates to a nuclear reactor core for the application of said method.
In the more exact terms, the present invention is intended to ensure that the heat flux released by a nuclear reactor core in the radial directions and in its axial direction is made as uniform as possible.
In order to gain a clearer understanding of the problem, reference will be made to FIG. 1 of the accompanying drawings in which is shown diagrammatically the curve of distribution of neutron flux within a nuclear reactor core.
More precisely, said core is concerned with the case in which the enrichment of the fissile material constituting the reactor core is homogeneous.
In this figure, the neutron flux is plotted as ordinates along the axis Oy and the distance from the axis of the reactor core is plotted on the axis of abscissae Ox. It is considered in this figure that the reactor core has symmetry of revolution about its axis Oy. Said neutron flux is found to decrease substantially from the axis to the periphery of the reactor core. The zone a represents the fissile region of the core and the zone c represents the fertile blanket. As is already known, this phenomenon is due to the fact that fast neutrons exhibit a strong tendency to escape at the periphery of the reactor core.
The curve of evolution of heat has substantially the same shape in the case in which the enrichment is the same throughout the reactor core.
It is clearly an advantage to obtain a flatter curve of evolution of heat.
FIG. 2 shows a known method employed precisely for flattening the curve which is representative of the evolution of heat. In this figure, the reactor core is shown diagrammatically in vertical half-section.
The axis Ox represents the distance with respect to the vertical axis Oy of the reactor core. In this embodiment, the central portion of the reactor core a is constituted by fissile material which has a degree of enrichment E.sub.1. The central zone a is surrounded by an annular peripheral zone having a thickness b. In this zone, the fissile material has a degree of enrichment E.sub.2 which is higher than E.sub.1. Finally and in accordance with known practice in breeder reactors, provision is made for a second annulus having a thickness c which constitutes the radial fertile blanket.
There is also shown in this drawing the curve I which gives the neutron flux as a relative value, that is, as a percentage of the maximum flux along the axis of the reactor core.
The full-line curve I just mentioned is a continuous curve.
The higher degree of enrichment within the zone having a thickness b simply has the effect of slowing-down the fall in neutron flux without, however, raising it to any considerable extent. This change is indicated by the point of inflexation A.
The evolution of heat is represented by the broken-line curve II. It is apparent that this non-continuous curve is constituted by three portions. With the scales adopted as ordinates, the first portion II.sub.a follows the curve which gives the neutron flux whereas in the first peripheral zone, the portion of curve II.sub.b is located distinctly above the neutron-flux curve I, this being clearly due to the increased degree of enrichment of the fissile material in this zone. Finally, the third portion II.sub.c corresponds to a very slight evolution of heat within the radial fertile blanket as a result of a very weak enrichment (natural uranium or more generally depleted uranium).
It is thereof readily apparent that this core structure permits a certain flattening of the radial curve of evolution of heat.
Moreover, there was described in U.S. Pat. No. 3,932,217 granted Jan. 13, 1976 to Pierre Charles Cachera and assigned to Electricite de France, a method for achieving enhanced safety of a fast reactor which consists in employing safety fuel elements in which is formed a central passage or axial flow duct of large diameter on the order of 10/15 of the diameter of the fuel element and having a sufficient diameter to ensure that the fuel which may be caused to melt within the central portion of the element under the action of an accidental power excursion is capable of flowing by gravity rapidly into the lower blanket placed as a catchpot without impairing the fuel can. Each safety element has a stack of fissile fuel pellets in the can with each pellet having this axial flow duct and also having a lower member of refractory material with an axial duct at the top of substantially the same diameter as the duct in the fissile portion.
This arrangement makes it possible to ensure enhanced reactor safety since there is obtained at the time of an accidental power excursion of reduction of the quantity of fissile material within the central region of the core in which the neutron flux has the highest intensity, thus resulting in a decrease in reactivity which automatically puts an end to said power excursion.